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Science and Technology of Nuclear Installations
Journal Prestige (SJR): 0.522  Citation Impact (citeScore): 1 Number of Followers: 3  Open Access journalISSN (Print) 1687-6075 - ISSN (Online) 1687-6083 Published by Hindawi  [339 journals]
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- Acceptable Level of Acceptance and the Affecting Factors: What Is the
Acceptable Public Acceptance of Building a Nuclear Power Plant
Abstract: This research determines the Acceptable Level of Acceptance (ALA) based on the countries with active Nuclear Power Plant (NPP). The ALA is a particular value of public acceptance of NPP, indicating public support and participation in the program. If the public acceptance level is lower than the ALA, then the probability of public resistance against the program is relatively high and would harm the NPP. There is no correlation between the number of populations. This research uses four categories to classify public acceptance: (1) low, (2) moderate, (3) high, and (4) very high. Based on these categories, this research suggests that the moderate ALA is 27.5% of the acceptance level.
PubDate: Sat, 11 Mar 2023 08:20:02 +000
- Modelling and Validation of CANDU Shim Operation Using Coupled TRACE/PARCS
with Regulating System Response
Abstract: In CANDU reactors, shim operation is used when the online refuelling capability becomes temporarily unavailable. Adjuster rods, normally in-core to provide flux flattening, are withdrawn in sequence to provide additional reactivity as the fuel is depleted. In a CANDU 900 reactor, up to three of the eight adjuster banks may be withdrawn, with the power derated accordingly. In this study, the shim operation was modelled using a combination of TRACE_Mac1.1, PARCS_Mac1.1, and scripts modelling the reactor regulating system, all running as a single coupled simulation. A driver script simulated the operation as a sequence of steady-state, depletion, and transient models. The results were compared to operational data from a nuclear power plant, evaluating the key figures of merit. The simulation was extended beyond the operational data by reducing the power to 59% FP and withdrawing the remaining adjusters, to observe the behaviour of the simulated reactor for a deeper reactivity-driven transient. Sensitivity cases, including adjuster rod depletion and nuclear data uncertainty, were also evaluated. This study was able to successfully reproduce the general results of the shim operation. Some discrepancies were observed between the simulation and dataset for the duration of the shim, particularly for the one bank out phase of the shim. Several potential causes for the early phase behaviour were identified. When the simulation was extended, the model predicted that a power reduction below 60% FP would lead to xenon poison out when the adjusters were depleted, with the timing sensitive to the adjuster depletion. Nodalisation of the PARCS model also had a significant impact, due to the effect on adjuster nodalisation and its area-of-effect with respect to the actual adjuster locations. Nuclear data uncertainty had a lesser but still noticeable effect. Other parameters, such as the distribution of fuel burnups in the core, only had a small effect on the shim operation.
PubDate: Wed, 08 Mar 2023 11:35:00 +000
- Steam-Jet Evaluation for Predicting Leakage Behavior and Interpretation of
Experimental Verification
Abstract: Owing to pipe thinning, fatigue damage, and aging, pipes, valves, and devices installed in the primary and secondary systems of nuclear power plants may leak high-temperature/high-pressure reactor coolant. Thus, a system must be developed to determine if the leakage is exceeding the operating limit of the nuclear power plant, thereby mitigating any loss of life or economic loss in such cases. In this study, a validated numerical analysis method was established to initially simulate the leakage behavior and subsequently to evaluate the small amount of leakage in the compartment. For this purpose, a vapor-jet collision test in the compartment and a vapor-jet test in the pipe were performed; numerical analysis was conducted, and comparative analysis was performed to verify the validity of the established method. The evaluation results suggested that the proposed numerical analysis method could optimally simulate the flow characteristics of the steam jet. Notably, compared to the existing evaluation method, the proposed approach simulated a more detailed behavior of the jet formed at the leakage point. In future research, the results of this study (data) will be used to inform the design of the second phase of the leak-capture system and will be served as the foundation for a performance-optimization study on the capture system.
PubDate: Tue, 07 Mar 2023 14:05:01 +000
- Assessing the Impact of Common Cause Failures on Site Risk within Level 1
Multi-Unit PSA
Abstract: Common cause failures (CCFs) may lead to the simultaneous unavailability or failure of numerous components in the nuclear power plant because of the existence of a shared cause when an initiating event disrupts the normal functioning of nuclear power plants. The presence of common cause failures (intra-unit and inter-unit) can be recognized in a multi-unit probabilistic safety assessment (MUPSA) as a crucial dependency factor that can influence accident scenarios and the core damage frequency (CDF), as CCF may affect the availability and proper operation of mitigating systems. Since such failures are likely to significantly undermine the benefits of the concept of redundancy in nuclear power plant systems, it is necessary to identify the CCFs that contribute to the core damage in a multi-unit site and analyse their overall quantitative magnitude and qualitative proportions. In this study, a twin-unit generic pressurized water reactor (PWR) nuclear plant is modeled using the AIMS-PSA software. For the loss-of-offsite-power (LOOP) and station blackout (SBO) events, the site CDF was calculated, and the cut-sets produced by this quantification were examined for the modeled CCF basic events in the fault trees. The quantitative and qualitative contributions of the CCFs to the frequency of site core damage were examined. CCFs in the modeled fault trees contributed to 4.58% to the site CDF of the combined LOOP followed by SBO event. In the LOOP event alone that leads to core damage, the CCF contributed 4.58% to the site CDF while CCFs contributed 17.19% to the site CDF in the SBO event alone that leads to core damage. With CCF events considered in the modeling process, the site CDF estimated with CCF events increased by 7.53% in the combined LOOP followed by SBO event. In the LOOP event alone that leads to core damage, inclusion of CCF events in the modeling increased the site CDF by 7.42%. A 15.66% increase in site CDF was recorded in the SBO event alone that leads to core damage as compared to modeling without CCF events. The results show how crucial the common cause failure contribution is to site CDF. The safety of the nuclear plant at a site is impacted by an increase in site CDF when common cause failures are considered. The various CCF fundamental event compositions and their percentage contributions were explicitly examined by the minimal cut-sets which leads to core damage in the units. In conclusion, this study’s findings can help us better understand how CCFs increase multi-unit site risk and can also act as a starting point for future studies on the qualitative and quantitative categorizations of CCF effects within MUPSA.
PubDate: Fri, 03 Mar 2023 11:35:02 +000
- Prediction of Automatic Scram during Abnormal Conditions of Nuclear Power
Plants Based on Long Short-Term Memory (LSTM) and Dropout
Abstract: A deep-learning model was proposed for predicting the remaining time to automatic scram during abnormal conditions of nuclear power plants (NPPs) based on long short-term memory (LSTM) and dropout. The proposed model was trained by simulated condition data of abnormal conditions; the input of the model was the deviation of the monitoring parameters from the normal operating state, and the output was the remaining time from the current moment to the upcoming reactor trip. The predicted remaining time to the reactor trip decreases with the development of abnormal conditions; thus, the output of the proposed model generates a predicted countdown to the reactor trip. The proposed prediction model showed better prediction performance than the Elman neural network model in the experiments but encountered an overfitting problem for testing data containing noise. Therefore, dropout was applied to further improve the generalization ability of the prediction model based on LSTM. The proposed automatic scram prediction model can provide NPP operators with an alert to the automatic scram during abnormal conditions.
PubDate: Fri, 03 Mar 2023 07:50:01 +000
- Simulations of Core Damage Progression for TMI-2 Severe Accident Using
CINEMA Computer Code
Abstract: As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fission product analysis in the vessel and the containment, respectively. The severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. This analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. The CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. The CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. The CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively.
PubDate: Mon, 27 Feb 2023 11:20:00 +000
- Effectiveness of Serpentine Concrete as Shielding Material for Neutron
Source Facility Using Monte Carlo Code
Abstract: In recent years, much attention has been dedicated to finding techniques to reduce exposure doses. This work examines the effectiveness of using serpentine concrete to shield a neutron source using a 241Am-Be neutron source facility at the National Nuclear Research Institute (NNRI) as a case study. The results obtained for both neutrons and gamma indicate that serpentine concrete provides better shielding as compared to ordinary concrete. At a distance of 100 cm from the Am-Be source, when shielded with serpentine concrete, it was found that personnel will receive an average gamma dose of 4.395.395 ± 0.122 μSv/h while a dose of 10.399 ± 0.083 μSv/h will be received for ordinary concrete shield. The average neutron dose equivalent at 100 cm, for ordinary concrete and serpentine concrete were 32.189 ± 0.277 and 9.276 ± 0.505, respectively. All dose equivalents obtained were also within internationally accepted limits.
PubDate: Sat, 18 Feb 2023 07:35:02 +000
- Development of an Integrated Human Error Simulation Model in Nuclear Power
Plant Decommissioning Activities
Abstract: In this study, an integrated human error simulation model in nuclear power plant (NPP) decommissioning activities (HEISM-DA) that can integrate and manage various factors affecting human errors is developed. In the HEISM-DA, an error probability input method suitable for the characteristics of each performance shaping factors (PSFs) was presented. Because each PSF has different importance on human error, the relative importance of decommissioning PSF Levels 1 and 2 and influential factors is considered. A multiplier was selected for each PSF and then used for human error evaluation. To calculate the human error probability (HEP) for the NPP decommissioning activity, the relationship between each PSF is identified and linked to develop a human error evaluation model. Using the HEISM-DA, HEP for reactor pressure vessel internal cutting work is evaluated based on the experience data. HEP is calculated to be approximately 1%. As a result of HEP calculation, it is found that the “operation” factor has a significant influence on the HEP of NPP decommissioning activities. Therefore, if the dismantling work is conducted by supervising the “operation” factors in a detailed and systematic approach, it is believed that the HEP will be reduced as other factors are also affected.
PubDate: Thu, 12 Jan 2023 09:20:01 +000
- Application of EDG AOT Extension Based on the Risk-Informed Method in NPPs
Abstract: At present, the allowed outage time (AOT) of an M310 unit emergency diesel generator (EDG) was 3 days, which can be extended to 14 days through replacement of additional diesel units; although it provides a certain online maintenance time, it cannot meet the needs of ten-years overhauls. In order to avoid stopping the reactor for maintenance of NPP due to insufficient of EDG AOT, based on risk-informed method analysis feasibility of extending AOT for EDG to 28 days, we quantitatively calculate the impact of extension of AOT on risk level of nuclear power plants (NPPs). Analysis shows that extension of EDG AOT to 28 days has less impact on NPPs, and safety of NPPs can be further ensured through temporary risk control measures, so the extension of AOT to 28 days is acceptable. By using risk-informed technology to extend AOT for EDG, unnecessary shutdown and maintenance is avoided and the economy of NPPs and flexibility of maintenance work arrangement is greatly improved while ensuring safety, which is of great significance to operation and maintenance of NPPs.
PubDate: Thu, 05 Jan 2023 12:05:02 +000
- High-Temperature Corrosion Behavior of Incoloy 800H Alloy in the Impure
Helium Environment
Abstract: The helium coolant in the primary circuit of the high-temperature gas-cooled reactor (HTGR) contains traces of impurities, which can induce the corrosion of superalloys when exposed to elevated temperatures. The superalloy damage caused by the corrosion could threaten the safe operation of the reactor. In this work, the corrosion behavior of a representative superalloy (chromium-rich iron base alloy Incoloy 800H) was investigated under the impure helium at different typical temperatures of HTGR. An experimental setup developed for studying the high-temperature corrosion of superalloys was used to investigate the chemical reactions and corrosion behaviors of Incoloy 800H. It was found that CO2 is an important oxygen source in the reaction with chromium, and CO is released as the product. In addition, the observation and computation of the critical temperature (TC) of the reaction between CO2 and carbon in the alloy show that TC is much lower than that (TA) of the microclimate reaction, which indicates that CO2 can protect the scale from destruction. Furthermore, the slight decarbonization of the alloy was found above TC. Also, a model developed by the thermodynamic analysis was proposed to explain the mechanism of slight decarbonization and predict the critical temperature when the CO2-C reaction occurs. This work presents a guideline for protecting the oxide scale of superalloys used in HTGR.
PubDate: Fri, 16 Dec 2022 07:50:00 +000
- Tritium Breeding Performance Analysis of HCLL Blanket Fusion Reactor
Employing Vanadium Alloy (V-5Cr-5Ti) as First Wall Material
Abstract: Neutronic analysis in the HCLL blanket module has been established, and the calculation was performed by the ITER team, including the first wall (FW). In this study, seven materials have been investigated for FW material by considering characteristics such as high neutron fluence capability, low degradation, under irradiation, and high compatibility for blanket material. A three-dimensional configuration simulated in MCNP5 program codes was performed to investigate the neutronic performance and radiation damage effect. Employing seven candidates are vanadium carbide (VC), titanium carbide (TiC), vanadium alloy (V-5Cr-5Ti), graphite (C), tungsten alloy (W-CuCrZr), ceramic alloy (SiC), and HT-9 to study optimization of FW materials configurated in the HCLL blanket module. This novelty study concludes that vanadium alloy (V-5Cr-5Ti) is becoming a promising material candidate. This alloy has the highest number of neutronic performing for 1.27 TBR and 1.26 in multiplication energy factor in all investigations. Meanwhile, the amount of atomic displacement, hydrogen, and helium production are around 22.31 appm, 765.55 appm, and 281.57 appm, respectively. Even though vanadium alloy has a reasonably high radiation damage effect, it is still tolerable compared to several thresholds of DPA. So, it is considered excellent material for FW. Nevertheless, this alloy can replace after 13.45 years for radiation damage.
PubDate: Thu, 15 Dec 2022 10:35:01 +000
- A Data-Driven Fault Prediction Method for Nuclear Power Systems Based on
End-to-End Deep Learning Framework
Abstract: With the increase in system complexity and operational performance requirements, nuclear energy systems are developing in the direction of intelligence and unmanned, which also requires a higher demand for its safety so that intelligent fault diagnosis and prediction have become a technology that nuclear power plants need to develop at present. At the same time, due to the rapid development of deep learning technology, it has become a meaningful development direction to predict the fault state of nuclear power plants within the framework of supervised deep learning. Usually, the network structure model used in fault diagnosis and prediction requires professional design, which may cost a lot of time and make it difficult to achieve optimal results. For this purpose, we present an end-to-end deep network for nuclear power system prediction (EDN-NPSP), which can automatically mine the transient features of various detection data in the NPS at the current moment through heterogeneous convolution kernels that can increase the receptive field and then predict the feature evolution results of the NPS in the future through a special deep CNN. The results provide an assessment of the future state of NPS. Based on EDN-NPSP presented in this work, we can avoid complicated manual feature extraction and provide the predicted state directly and rapidly. It will provide operators with useful prediction information and enhance the nuclear energy system fault prediction capabilities.
PubDate: Tue, 29 Nov 2022 06:05:00 +000
- Nuclear Power Sustainability Path for China from the Perspective of
Operations
Abstract: Nuclear power, as a low-carbon, stable, and efficient energy, plays an important role in replacing fossil fuels in the development of a globally sustainable energy system. However, nuclear power has deviated from the path to achieve the Sustainable Development Goals of the United Nations. The path of sustainable nuclear power for China was proposed based on an analysis of the development of global nuclear power and the situation in China, using advanced operation concepts and intelligent collaboration technology to change the labor-centered operation mode. It serves as a model for other countries with a labor-centered nuclear power operation mode and an aging society seeking to achieve carbon neutrality through the use of nuclear power around the world.
PubDate: Mon, 28 Nov 2022 05:35:02 +000
- Investigation of Early Corrosion Behavior of Canister Candidate Materials
in Oxic Groundwater by the EQCM Method
Abstract: This study investigated the corrosion mass changes of canister candidate materials (Cu, Ni, Ti, SS304) in an oxic groundwater solution using the electrochemical quartz crystal microbalance method in order to estimate corrosion thickness. The materials were immersed in naturally aerated groundwater with and without the addition of chloride ions to observe the mass changes as well as the open-circuit potential (corrosion potential). In the oxic groundwater solution, Ni, Ti, and SS304 exhibited negligible mass changes, indicating their insusceptibility to general corrosion. In contrast, the Cu electrode exhibited a relatively significant mass change (63.8 ng/cm2 for 60 h), and the maximum corrosion thickness was estimated to be approximately 0.1 μm/yr. In the presence of chloride ions, the Ni and Ti electrodes did not reveal demonstrate any significant changes, whereas the SS304 electrode was slightly increased compared to an absence of chloride ions. A lower mass change occurred when the Cu electrode was immersed in the chloride-containing groundwater solution compared with the absence of chlorides because the dissolution of Cu as was involved in Cu2O formation.
PubDate: Thu, 17 Nov 2022 12:20:01 +000
- Development of an MPS Code for Corium Behavior Analysis: 3D Alloy Melting
Abstract: The moving particle semi-implicit (MPS) method as a Lagrangian method is attracting increasing attention in severe accident analysis. In this paper, we developed an MPS code for the corium behavior analysis with several additional models added: an improved heat transfer model to improve the calculation between different materials, an enthalpy-based viscosity model to realize a smooth transition of viscosity at the solid-liquid interface, and a surface tension model for better simulation of surface shape. Validation of the developed simulation approach is carried out on a classical water column collapse example. The development of the heat transfer model is validated by the example of a one-dimensional semi-infinite plate. A comprehensive example of the melting of “Wood’s alloy” is carried out to verify the capacity of MPS method in the simulation of melting and expansion. The simulation results are in good agreement with the experimental results, which indicates that MPS method promises well in the field of severe accidents.
PubDate: Thu, 03 Nov 2022 12:50:01 +000
- Feasibility Study on the Initial Kartini Reactor Core Using Plate Type
Fuel Elements
Abstract: The plate type fuel element conversion is proposed to solve a supply problem of TRIGA standard rod type fresh fuel in the long term and to extend the lifetime by reducing the dependence of buying imported elements. The plate type fuel is an alternative since the Indonesian industry has been able to produce such fuel elements. The change of core configuration is expected to improve the reactor performance for irradiation facilities and fuel element lifetime. The SRAC2006 is used to perform neutronic calculations while the nuclear fuel lifetime is calculated by SWAT. This study begins with performing a core properties comparison of UZrH1.6 as the current fuel material and U3Si2-Al as the fuel material candidate. The results show that the Kartini reactor core is possible to load U3Si2-Al as the fuel material and makes higher excess reactivity compared to the current fuel material. Furthermore, U3Si2-Al in the plate type element geometry is variedly arranged in the new reactor core configuration to optimize the neutronic core parameters. The new core configuration is composed of 10 standard fuel elements, 4 fuel control elements, and the graphite material baffle that is located between the active core and annular reflector for serves as an additional reflector. The configuration produced sufficient core excess reactivity and adequate shutdown margin. It also produced negative temperature feedback reactivity and power peaking factor that fulfilled the safety requirements. Improvement of new reactor core performance was obtained by more irradiation facilities, higher thermal neutron flux, and longer maximum estimated burn up compared to the current core configuration.
PubDate: Fri, 14 Oct 2022 03:50:02 +000
- The Study of Dosimetric Characteristics of the XHA600D Medical Linear
Accelerator Based on a Monte Carlo Code
Abstract: By investigating the influence of initial electrons on dosimetric characteristics, reasonable incident electron parameters for the nominal 6 MV photon beam of the XHA600D accelerator are finally established, i.e., a 6 MeV monoenergetic electron beam with a radial intensity FWHM of 2.5 mm and an angular divergency of 0.15°. Based on reasonable initial parameters, Percentage Depth Doses (PDDs), Off-Axis Ratios (OARs), total scatter factors, beam qualities, and penumbra widths of both flatteningfilter (FF) and flattening-filter-free (FFF) beams for fields ranging from 4 × 4 to 30 × 30 cm2 are simulated systematically with EGSnrc codes. Not only the simulated dosimetric properties are in excellent agreement with the measurements, but also the dosimetric discrepancies between FF and FFF beams are consistent with the laws of previous studies on other accelerators. Therefore, reasonable incident electron parameters are able to accurately verify the performance of the XHA600D accelerator and can be used for further dosimetry research.
PubDate: Thu, 29 Sep 2022 06:05:02 +000
- Research on Radionuclide Diffusion Mechanism in the Ocean and Emergency
Response under Oceanic Radioactive Events
Abstract: On March 11, 2011, a serious radionuclide leakage accident occurred at Fukushima Daiichi nuclear power plant, and a large number of radionuclides were released, causing serious pollution to the ocean environment. On August 25, 2021, Japan announced the overall plan for the discharge of radioactive sewage from the Fukushima Daiichi nuclear power plant into the ocean, and the discharge will begin around the spring of 2023. All operational and under-construction nuclear power plants in China are distributed in coastal areas presently. In case of a nuclear leakage accident, radionuclides will diffuse through the ocean and pollute the ecological environment. The study of radionuclide diffusion mechanism in the ocean and emergency response plays an important role in accident mitigation under oceanic radioactive events. A radionuclide diffusion model in the ocean was established and the radionuclide diffusion mechanism in the ocean was analyzed. And then a prediction and monitoring system of radionuclide diffusion in the ocean was proposed. The results show that the short-term radionuclide diffusion is mainly influenced by the source term, flow field and decay of 131I, and the degree of influence decreases in turn. On the whole, influences of the flow field and 131I decay are weakened during the long-term diffusion. At the same time, the influence of 137Cs decay begins to be obvious and the influence of suspended matter is increasing. The influence of ocean organisms is always small. Problems of scientific prediction and protection were analyzed, and the emergency response scheme was given. It is of great significance to improve the capacity of emergency response for oceanic radioactive events.
PubDate: Sat, 24 Sep 2022 12:50:01 +000
- Research and Analysis on Assembly Methodology for the Fusion Blanket
System
Abstract: In the development process of the blanket system, assembly design is quite important, and suitable systematic methodologies are required. As we know, the CFETR machine and the associated fusion components are not usually mass-produced large-scale products, but highly customized machines which are still in the design phase. Appropriate assembly methodology plays an important role in fulfilling the function of the fusion machine. This paper has investigated some universal assembly methods for similar complex products. Two preferred methods of design-for-assembly (DFA) and product-process hierarchical modeling (PPHM) have been analyzed and improved taking the fusion blanket system as a study case. The overall process of the blanket system was studied including the stages of design, assembly, and overview of the blanket system hierarchy structure. The two newly proposed methods aim to clarify a probably feasible assembly method for the blanket system, though it is still in the engineering design stage. Case studies of the two favorable assembly methods can be good references to demonstrate and analyze the advantages of DFA and PPHM for decision-making in each product development phase.
PubDate: Wed, 21 Sep 2022 07:35:03 +000
- Experimental Study on Unsteady Radon Exhalation from the Overburden Layer
of the Uranium Mill Tailings Pond under Rainfall
Abstract: In order to find out radon reduction performance of the overburden layer on uranium mill tailings (UMTs) pond beach surface after rainfall, the rainfall simulation experiment of the overburden layer was carried out with the self-developed equipment. Based on the radon migration model of the overburden layer on the UMTs pond beach surface, the change rule of radon exhalation in four types of compactness of the overburden layer within 120 hours after rainfall was studied, and the corresponding moisture content was also analyzed. The results show that the radon concentration in the overburden layer of UMTs increases nonlinearly; the dynamic change in moisture content of the overburden layer on the beach surface leads to the unsteady radon exhalation. The variation of radon exhalation shows three stages: increase, linear decrease, and stability tendency. After rainfall, radon exhalation rate increases due to water vapor and there is free radon seepage in pores. With the decrease of free radon production rate, radon exhalation rate gradually decreases until it reaches stability again. When the thickness of the overburden layer reduces, the porosity decreases with the increase in compactness of the overburden layer. While the decrease in radon reduction is more obvious, the less time it takes for radon exhalation to vary from unstable to stable overburden after rainfall.
PubDate: Mon, 19 Sep 2022 10:35:04 +000
- Stress and Strain State Analysis of Crack Front in Dissimilar Metal Welded
Joints with Dual Field of Mechanical Heterogeneity and Residual Stress
Abstract: The micro-mechanical state at the crack front is one of the key factors affecting the stress corrosion cracking (SCC) growth behavior. The mechanical heterogeneity and residual stress in the dissimilar metal welded joint (DMWJ) induce the micro-mechanical state at the crack front to become more complex. The sandwich model and dual-field model of the DMWJ with inner surface axial crack were established in this study. The stress and strain states at the crack front with different crack locations and lengths under the interaction of the mechanical property and the residual stress were investigated. The results show that a more accurate evaluation of stress and strain states can be obtained when using the dual-field model to describe the material mechanical property and residual stress of the DMWJ. The sandwich model overestimates the crack driving force including the stress and strain at the crack front. The tensile stress in the middle of shallow cracks is smaller than that at both ends, while the tensile stress in the middle of deep crack is larger than that at both ends. The variation trend of the tensile stress and normal strain at the crack apex is basically the same as that of the residual stress with the crack depth. However, there is almost no normal plastic strain in the initial stage of crack propagation due to the small residual stress in the initial stage.
PubDate: Sat, 17 Sep 2022 06:35:02 +000
- Experimental and Numerical Studies of AP1000 Shield Building considering
Fluid-Structure Interaction
Abstract: The gravity cooling water tank is a remarkable structural feature of third-generation pressurized water reactor nuclear power plant. To investigate the influence of fluid-structure interaction (FSI) on the seismic response of the structure, this study designed two 1 : 50 simplified models of the AP1000 shield building. A series of shaking table tests were conducted to study the seismic responses with and without FSI effect. The natural frequency, acceleration, strain, and hydrodynamic pressure of the two models were analyzed, and the seismic reduction effect of the water tank was evaluated. Moreover, the test data were compared with the results of numerical analysis using the ABAQUS software. The results show that the presence of water and the sloshing of water reduce the natural frequency and seismic response of the model structure. Thus, the gravity cooling water tank has a certain seismic reduction effect. The simplified model of water sloshing can be used to analyze the seismic response of the shield building.
PubDate: Fri, 02 Sep 2022 08:05:01 +000
- Machine Learning-Based Approach for Hydrogen Economic Evaluation of Small
Modular Reactors
Abstract: In this study, we evaluate hydrogen production costs using small modular reactors (SMRs). Furthermore, we employ a machine learning-based approach to predict important parameters that affect the hydrogen production cost. Additionally, we use a hydrogen economic evaluation program to calculate the hydrogen production cost when using the two types of SMRs: system-integrated modular advanced reactor (SMART) developed by the Korea Atomic Energy Research Institute (KAERI) and NuScale power module™ (NPM) developed by the NuScale Power, LLC. Different storage and transportation means were selected to find the cheapest option. Using SMART, storing hydrogen in compressed gas and transporting it through pipes (CG-Pipe) is the best option, with an estimated cost of USD 2.77/kg. Other options when using SMART include storing in compressed gas and transporting with a vehicle (CG-Vehicle), with an estimated cost of USD 3.27/kg; storing by liquefaction and transporting with a vehicle (L-Vehicle), with an estimated cost of USD 3.31/kg; and storing in metal hydrides and transporting with a vehicle (MH-Vehicle), with an estimated cost of USD 6.97/kg. Using NPM, CG-Pipe is the cheapest option to generate hydrogen, with an estimated cost of USD 2.95/kg. Other options include CG-Vehicle (USD 3.35/kg), L-Vehicle (USD 3.42/kg), and MH-Vehicle (USD 7.04/kg). Hydrogen production using SMART is cheaper than using NPM. However, the observed difference between the hydrogen production costs using the two reactors was insignificant. We conclude that the optimal hydrogen production cost ranges from USD 3.27/kg (CG-Vehicle) to USD 3.42 (L-Vehicle). This conclusion is because the common hydrogen transportation means is with a vehicle. From a machine learning-based approach, we determine the important parameters that affect hydrogen production costs. The most important parameter is the heat consumption (MWth/unit) at hydrogen generation plants, and other parameters include electricity rating and heat for hydrogen generation plants.
PubDate: Thu, 01 Sep 2022 04:50:03 +000
- Numerical Study of Unsteady Pressure Fluctuation at Impeller Outlet of a
Centrifugal Pump
Abstract: Intense fluid-dynamic interaction at the impeller outlet strongly affects the unsteady flow and pressure stability within the centrifugal pump. In order to have a better understanding of the pressure fluctuation of centrifugal pumps, a numerical calculation is carried out by using the RNG k-epsilon turbulence model under various flow rates. The numerical calculation results are compared with the experimental results in order to verify the reliability of the calculation model. The amplitude and frequency distribution of pressure fluctuation at the impeller outlet is obtained and analyzed in the time and frequency domain. The research results show that the blade passing frequency is the dominant frequency of the pressure fluctuation. And the pressure fluctuation is a periodic fluctuation. As the flow rate decreases, the periodicity of the pressure fluctuation decreases. Besides, the amplitude and intensity of pressure fluctuation are closely related to flow rate and spatial location. At the low flow rate, the amplitude of pressure fluctuation in the time domain and frequency domain is enlarged greatly, especially near the tongue region. The pressure difference distribution on both sides of the blade surface is extremely uneven, and the pressure changes significantly.
PubDate: Tue, 30 Aug 2022 11:05:06 +000
- A Hybrid Method to Predict the Remaining Useful Life of Scroll Wheel of
Control Rod Drive Mechanism
Abstract: As one of the rotating components in the reluctance motor type control rod drive mechanism (CRDM), the life of the scroll wheel is closely related to the service life of the CRDM. In addition, the prediction of the remaining useful life of the scroll wheel helps to optimize the maintenance process of the CRDM. This paper proposes a hybrid method to predict its remaining useful life when the available degradation data are rare and the failure threshold cannot be accurately defined. First, the particle filtering algorithm, whose state transfer equation is established on the segmental damage physical model, is used to predict the degradation state of the scroll wheel. Second, the proportional hazard model for the relationship between the scroll wheel damage characteristics and reliability model is established to predict the remaining useful life of it. The proposed method focuses on the establishment of segmental damage physical model and the clustering analysis of damage characteristics extracted from vibration signals, which can be used to predict the remaining useful life of the scroll wheel. In addition, the results provide an opportunity for the condition-based preventive maintenance of the CRDM.
PubDate: Mon, 29 Aug 2022 12:20:02 +000
- Numerical Study of Natural Circulation Flow in Reactor Coolant System
during a Severe Accident
Abstract: The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.
PubDate: Mon, 29 Aug 2022 09:35:01 +000
- Debris Bed Self-Leveling Mechanism and Characteristics for Core Disruptive
Accident of Sodium-Cooled Fast Reactor: Review of Experimental and
Modeling Investigations
Abstract: Evaluations of the Core Disruptive Accident (CDA) are significantly important for safety analysis of Sodium-cooled Fast Reactor (SFR) despite the very low probability of occurrence for CDA. During the material-relocation phase in CDA of SFR, the molten materials are possibly released from the core region into subcooled sodium, subsequently forming the debris bed on the lower part of the reactor vessel after being quenched and fragmented. The accumulated high-temperature debris with decay heat can cause sodium coolant boiling, leading to the so-called “debris bed self-leveling behavior” during which the shape of the debris bed becomes flattered (leveling). It is important to investigate the debris bed self-leveling behavior due to its potential capacity to induce the transfer of debris and affect the ability of cooling and criticality of the debris bed. Thus, in recent years, valuable knowledge concerning the mechanism and characteristics of this behavior was accumulated through lots of experimental results and modeling developments. Aimed at providing a valuable guideline for future investigations on this issue, in this study, the past experimental and modeling investigations on debris bed self-leveling mechanism and characteristics are systematically summarized and reviewed, and some future remarks are also proposed to promote the progression of further research for SFR severe accident analysis.
PubDate: Mon, 22 Aug 2022 14:20:02 +000
- Investigation of Oxidation and Counter-Oxidation in a One-Quarter Circular
Geometry due to Shadow Corrosion
Abstract: To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity and shadow corrosion was examined under two different cathodes. This work also investigates the effect of current exchange density and the cathodic Tafel coefficient on the cathodic current. Using COMSOL Multiphysics 5.2, the Laplace equation was solved to obtain the electrostatic potential and current density distributions in the studied domain. When the distance d between the anode (Zry-2) and cathode (platinum/nickel) is ≤0.5 mm, it was observed that a uniform oxide layer of thickness 20 µm grew on the smooth internal surface of Zry-2 for corrosion lasting 1166 h. When d > 0.5 mm, the oxide thickness falls in a manner dictated by the degree of dissociation α of the electrolyte. At a cladding gap of 0.08 mm, a radiation-enhanced uniform corrosion rate of 2.405 10−1 mmpy was obtained for Zry-2. This value is 142 times greater than that obtained at room temperature in the absence of radiation. It was also observed that the corrosion rate falls at higher cladding gaps, and the rate of change depends on the degree of dissociation. Other phenomena such as the dynamics of shadow corrosion under varying electrode separation and electrolyte conductivities, as well as extensive evaluation of critical fuel cladding parameters, are presented in this work.
PubDate: Tue, 09 Aug 2022 07:05:02 +000
- Coincidence Summing Factor Calculation for Volumetric γ-ray Sources
Using Geant4 Simulation
Abstract: Geant4 simulation was applied to correct the coincidence summing (CS) effect in detecting a volumetric γ-ray sources, and this technique was applied to a152Eu standard sources. The radioactive sources were a liquid cylindrical, rectangular, and Marinelli beaker shapes of different volume for each one. Radionuclide track (RT) including coincidence summing and monoenergetic track without coincidence summing. The results obtained from two approaches compared with the experimental data and the modified KORSUM code for cylindrical γ-ray source. The comparison showed that the adopted method in this investigation is useful for coincidence summing corrections for a voluminous γ-ray sources.Moreover, this technique requires far less computation time than the techniques that depend on the calculation of total efficiency.
PubDate: Mon, 08 Aug 2022 09:05:01 +000
- Effect of Chloride Ions on Electrochemical Behavior of Canister Materials
Abstract: Various canister candidate materials (SS316L, Ti-Gr.2, Alloy 22, and Cu) were studied using groundwater at the Korea Atomic Energy Research Institute (KAERI) underground research tunnel (KURT), with the addition of chloride ions using different electrochemical techniques. The corrosion potential and corrosion current of test materials were obtained by the polarization measurement. The polarization measurements revealed that the addition of chloride ions was detrimental to the SS316L and Cu materials by increasing corrosion current, which is an indicator of corrosion rate. Impedance measurements and fitting analysis showed that the corrosion resistance of Cu was more than 10 times lower than that of other materials in the KURT groundwater solution containing 0.1 M of chloride ions.
PubDate: Wed, 03 Aug 2022 03:50:01 +000
